A reactor pressure vessel (RPV) of a boiling water reactor (BWR) typically has a generally cylindrical shape and is closed at both ends, e.g., by a bottom head and a removable top head. A top guide typically is spaced above a core plate within the RPV. A core shroud, or shroud, typically surrounds the core and is supported by a shroud support structure. Particularly, the shroud has a generally cylindrical shape and surrounds the both the core plate and the top guide. The top guide includes several openings, and fuel bundles are inserted through the openings and are supported by the core plate.
A plurality of openings are formed in the RPV so that valves, nozzles, and pipes can extend within the RPV. For example, water enters the RPV through an inlet nozzle in the RPV sidewall. Similarly, control rod drive housings, e.g., tubes, are inserted through the bottom head and extend into the RPV. Moreover, a known RPV typically includes several openings for pressure instrument nozzles, and at least one opening for a drain nozzle. Since these components, e.g., control rod drive housings and nozzles, penetrate into the RPV, the components sometimes are referred to in the art as penetration tubes.
Mechanical wear, thermal cracking, and stress corrosion cracking (SCC) are known phenomena occurring in reactor components, such as structural members, piping, nozzles, valves, fasteners, and welds, exposed to high temperature water. The reactor components are subject to a variety of stresses associated with, e.g., differences in thermal expansion, the operating pressure needed for the containment of the reactor cooling water, and other sources such as residual stresses from welding, cold working and other inhomogeneous metal treatments. In addition, water chemistry, welding, heat treatment and radiation can increase the susceptibility of metal in a component to SCC.
To alleviate the effects of SCC, it typically is desirable to modify or upgrade existing pipes and valves within an RPV. When modifying or upgrading, for example, a valve, it is desirable to accurately map, or measure, the existing valve bore. Specifically, it is desirable to measure the roundness, diameter, sealing surface condition, concentricity and length of the valve bore before upgrading the valve.
One known method of measuring pipe and valve bores requires a person to manually measure such bores. Particularly, an operator typically must enter the bore area to measure the bore with a hand-held micrometer. This "hands on" approach, however, is both time consuming and difficult. Particularly, depending on the size and location of the pipe to be measured, it often is difficult for a person to accurately map the entire length of the bore. Furthermore, this method often is not practical because of the radiation levels and contamination levels associated with the location of the pipes and valves.
It would be desirable, therefore, to provide an apparatus for facilitating measuring pipe and valve bores in a nuclear reactor more easily and quickly than by known methods. It also would be desirable to provide such an apparatus to reduce operator exposure to radiation which measuring pipe and valve bores. It further would be desirable to provide such an apparatus which is inexpensive and easy to operate.